Strain Rate Evaluation of Some Typical Nuclear Power Plant Components During Plant Operation

The dynamic aspects of loading conditions for reactor internals, piping and the like, are thought to play important roles in the initiation of failures due, for example, to stress corrosion cracking (SCC) and fatigue. Some reports show that a strain rate on the order of 10−7 s−1 most affects susceptibility to SCC in the BWR reactor water environment. Environmental fatigue, which exhibits a shorter fatigue life in reactor water than that in air, is considered to have a remarkable correlation with strain rate and its affect on fatigue life. Despite its significant affect on SCC and fatigue, the actual strain rate of components is not known and practical evaluation methods have not been developed; consequently, such failure modes as SCC and fatigue are not evaluated in design. For this paper, strain rates induced by dynamic loading during such operations as plant start-up were calculated at typical points, such as reactor internals, piping and so on. The finite element method was applied to calculate the strain history of each point, and the strain rate was evaluated. The strain rate evaluation results clearly demonstrated that thermal transients provide greater peak strain rate values than pressure transients. Strain rates on the order of 10−7 s−1 were obtained for most points of major components during such thermal transients as plant start-ups. The major factors determining the strain rate magnitude were discussed, based on the calculation results. It was shown that the rate of temperature rise was the most important parameter, because it exhibited much larger sensitivity than the other parameters on the strain rate and could be controlled by plant operation procedures. In addition, a simple strain rate evaluation method based on Green’s function was developed for a specific point with a given design condition.

Probabilistic Fracture Analysis for Boiling Water Reactor Pressure Vessels Subjected to Low Temperature Over-Pressure Event

With the development of probabilistic fracture mechanics (PFM) methods in recent years, the risk-informed approach has gradually been used to evaluate the structural integrity and reliability of the reactor pressure vessels (RPV) in many countries. For boiling water reactor (BWR) pressure vessels, it has been demonstrated that it is not necessary to perform the inservice inspections of beltline circumferential welds to maintain the required safety margins because their probability of failure is orders of magnitude less than that of beltline vertical welds, thus may well reduce the associated substantial cost and person-rem exposure. In Taiwan, however, the inservice inspections of shell welds still have to be performed every ten years per ASME Boiler and Pressure Vessel Code, Section XI inspection requirements for a BWR type Chinshan nuclear power station. In this work, a very conservative PFM model of FAVOR code consistent with that USNRC used for regulation is built with the plant specific parameters concerning the beltline shell welds of RPVs of Chinshan nuclear power station. Meanwhile, a hypothetical transient of low temperature over-pressure (LTOP) event which challenges the BWR RPV integrity most severely is also assumed as the loading condition for conducting the PFM analyses. Further, the effects of performance of inservice inspection are also studied to determine the benefit of the costly inspection effort. The computed low probability of failure indicates that the analyzed RPVs can provide sufficient reliability even without performing any inservice inspection on the circumferential welds. It also indicates that performing the inservice inspections can not promote the compensating level of safety significantly. Present results can be regarded as the risk incremental factors compared with the safety regulation requirements on RPV degradation and also be helpful for the regulation of BWR plants in Taiwan.

Flamanville 3 EPR: How to Define the Pre-Service NDE Inspection Program

The last generation of PWR, the European Pressurizer Reactor, is being built both in Finland at Olkiluoto and in France at Flamanville where it will be the second unit in operation in two years. The conception of the reactor has no fundamental differences compared with the last French and German generations (i.e. N4 1450 MW and KONVOI 1300 MW). In fact the EPR is a synthesis of the best knowledge from both parts of Rhin river. Nevertheless, for what concerns the primary and secondary systems, few new issues have been implemented and they have important consequences for the pre-service inspection programme. First of it, main coolant lines and main components are declared break preclusion that means another approach for the in-service inspection and the requirements for pre-service inspection. A second one, is that the 1999 ministerial order will apply for the first time on a new unit. The inspection programme has to take it into account particularly in terms of NDE performance demonstration. In terms of structural integrity, the most important areas have been reviewed with fast fracture and fatigue criteria to determine the levels of NDE qualifications. This paper describes the important steps to reach the PSI and what is being developed in terms of NDE in relation to the structural integrity. Several examples are given to illustrate how EDF prepares the PSI of the EPR.

The Geometry of Steam Generator Tube and Its Relevance to the Occurrence of Stress Corrosion Cracking in Operating Nuclear Power Plants

Occurrences of a stress corrosion cracking in the steam generator tubes of operating nuclear power plants are closely related to the residual stress existing in the local region of a geometric change, that is, expansion transition, u-bend, ding, dent, bulge, etc. Therefore, information on the location, type and quantitative size of a geometric anomaly existing in a tube is a prerequisite to the activity of a non destructive inspection for an alert detection of an earlier crack and the prediction of a further crack evolution [1].

Dynamic Structural Analysis of the 9516 Transport Package

This paper presents the structural dynamic analysis of the 9516 package subjected to two hypothetical accidents in sequence: (1) the 30-ft drop test and (2) the puncture test. The analysis is performed with the general-purpose finite element analysis computer code, ABAQUS, using explicit integration. The 9516 package consists of a containment vessel placed inside a cask mounted inside a steel cage, which acts as the impact limiter and the personnel shield. The cask has a bolted closure that provides the confinement to the containment vessel. The closure bolts initially were modeled with one-dimensional elastic connector elements bearing only the axial loads, which was adequate for the analysis of the drop test. However, the closure lid experienced significant bending during the puncture test, implying bending load acting on the closure bolts. Therefore, the closure bolts were modeled by using solid elements in a refined model so that the response to the bending load during the puncture test can be simulated. The results of the analysis showed that the closure bolts experience significant bending during the puncture test. For model validation, a comparison between the analysis results and the test results for rigid body accelerations of the package is presented. The results of the analysis based on the refined model show that the maximum stress intensities in the closure bolts, containment vessel, and cask are lower than the corresponding allowable stresses specified in the American Society of Mechanical Engineering Boiler and Pressure Vessel Code.

Preshipment Leak Testing of 9977 Containment Vessels With the TM Electronics Solution Model S1A-L2-V Leak Tester

Pacific Northwest National Laboratory (PNNL) is replacing its 6M nuclear shipping fleet with 9977 shipping packages, which were designed by Savannah River National Laboratory (SRNL). The new packages require PNNL to perform a preshipment leak test on the lid seals of the containment vessel before the package is shipped on public roads. Savannah River National Laboratory (SRNL) developed a preshipment leak test using a TM Electronics Solution leak tester for PNNL. The Solution is an automatic vacuum leak tester that uses the Gas Pressure Rise leak test method to check the O-ring lid seals and the test port plug seal. The two tests take three minutes each to perform. Because the Solution is fully automatic, the leak test can be done by operators after a few hours of training. This paper describes the test equipment and the testing sequence.

Guide for Applying Codes and Standards for Safe Operation and Maintenance of Pressure Equipment in Refinery and Petrochemical Services

After pressure equipment (aka fixed or static equipment) is designed, fabricated, and constructed to ASME new construction codes and standards (C/S), it is delivered and placed in-service. After that the In-service Inspection (ISI) and Post-Construction C/S begin to govern. Within the ASME, the Post Construction Committee (PCC) produces and maintains the standards that govern equipment after it has been placed inservice. Within the API Standards Organization, the Subcommittee on Inspection (SCI) and the Corrosion and Materials Subcommittee (CMSC) produce and maintain most of the ISI standards and recommended practices that govern pressure equipment in the refining and chemical process industry. This paper shows how many of those ISI and PCC C/S are intended to work together to maintain the safety and reliability of pressure equipment and piping after it has been placed in service. This paper also highlights what’s new with many of these C/S that have been recently updated or newly published. Both the API and ASME use the rigorous, standardized consensus building process outlined by the American National Standards Institute (ANSI) for formulating and publishing their respective C/S. This paper will show how users of these ISI/PCC codes and standards are benefited by the application of the ANSI consensus process.

Weld Repair for Pressure Vessels Made From Cr-Mo Steels

Pressure vessels made from Cr-Mo steels are utilized for high temperature and high pressure services including hot hydrogen services. After long term operation, there are several past experiences of damages and/or degradation of materials such as temper embrittlement, creep embrittlement, hydrogen attack and hydrogen embrittlement. This paper summarizes typical damages/degradation and examples of weld repairs including special attention to development of weld repair procedure. The subject equipments are heavy wall petroleum pressure vessels made from Cr-Mo steel with austenitic stainless steel overlay cladding. Cracking could be prevented by controlling the repair welding process to reduce the hydrogen content at the interface. After repair welding, adequate post weld heat treatment (PWHT) has to be executed. Recently, repair welding has become an important aspect as part of post construction codes for pressure equipment to keep safe and long term continuous operation of the process plants because many of the plants have been operated for more than thirty years in Japan. Responding to the needs of petroleum and chemical industries, The Chemical Plant Welding Research Committee (CPWRC) of The Japan Welding Engineering Society (JWES) established the Pressure Equipment Repair Welding Subcommittee (PERW S/C) [1]. The S/C has developed optimum repair welding methods and procedures in the guideline on November 2009, with reference to the above investigation results. This paper also introduces the repair welding guideline for the pressure vessels made from Cr-Mo steels.

Regulatory Issues Associated With Shipment of Small Quantities of Radioactive Material

Following decertification of the ubiquitous and simple Department of Transpsortaion (DOT) 6M specification package, radioactive materials package Shippers have been faced with the need to use Certified Type B packagings. Many Department of Energy (DOE), commercial and academic programs have a need to ship small masses of radioactive material, where the identity of the material or radionuclides is know but the individual activity of some may not be known. For quantities which are small enough to be fissile exempt and have adequate shielding to ensure low radiation levels, these materials could be transported in a package which provides the required containment level. Because their Chalfant type containment vessels meet the American National Standards Institute (ANSI) N14.5 definition for leak-tight (≤ 1×10−7 ref cm3 air/sec), the 9975, 9977, and 9978 are capable of transporting contents requiring the highest standard of containment. The issues associated with certification of a high-integrity, general purpose package for shipping small quantities of such radioactive material are discussed and the logical basis for certification for such contents is described.

Aging Model for Cane Fiberboard Overpack in the 9975 Shipping Package

Many radioactive material shipping packages incorporate a cane fiberboard overpack for thermal insulation and impact resistance. Mechanical, thermal, and physical properties have been measured on cane fiberboard following thermal aging in various temperature/humidity environments. Several of the measured properties change significantly over time in the more severe environments, while other properties are relatively constant. Changes in each of the properties have been fit to a model to allow predictions of degradation under various storage scenarios. Additional data are being collected to provide for future refinements to the models.